Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 30127
Uncertainty Analysis of ROSA/LSTF Test on Pressurized Water Reactor Cold Leg Small-Break Loss-of-Coolant Accident without Scram

Authors: Takeshi Takeda

Abstract:

The author conducted post-test analysis with the RELAP5/MOD3.3 code for an experiment using the ROSA/LSTF (rig of safety assessment/large-scale test facility) that simulated a 1% cold leg small-break loss-of-coolant accident under the failure of scram in a pressurized water reactor. The LSTF test assumed total failure of high-pressure injection system of emergency core cooling system. In the LSTF test, natural circulation contributed to maintain core cooling effect for a relatively long time until core uncovery occurred. The post-test analysis result confirmed inadequate prediction of the primary coolant distribution. The author created the phenomena identification and ranking table (PIRT) for each component. The author investigated the influences of uncertain parameters determined by the PIRT on the cladding surface temperature at a certain time during core uncovery within the defined uncertain ranges.

Keywords: LSTF, LOCA, scram, RELAP5.

Digital Object Identifier (DOI): doi.org/10.5281/zenodo.2571872

Procedia APA BibTeX Chicago EndNote Harvard JSON MLA RIS XML ISO 690 PDF Downloads 238

References:


[1] USNRC, “Through-wall circumferential cracking of reactor pressure vessel head control rod drive mechanism penetration nozzles at Oconee nuclear station, Unit 3,” NRC Information Notice 2001-05, USNRC, Washington, DC, 2001.
[2] A. Cuadra, J. Ragusa, T. Downar, and K. Ivanov, “Analysis of a CRDM nozzle break LOCA without scram using the U.S. NRC coupled code TRAC-M/PARCS,” in Proc. of the 10th International Conference on Nuclear Engineering (ICONE-10), Arlington, USA, April 2002.
[3] The ROSA-V Group, “ROSA-V Large Scale Test Facility (LSTF) System Description for the Third and Fourth Simulated Fuel Assemblies,” JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.
[4] T. Takeda, H. Asaka, and H. Nakamura, “Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation,” Ann. Nucl. Energy, vol. 36, 2009, pp. 386–392.
[5] USNRC Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual,” NUREG/CR-5535/Rev 1, Information Systems Laboratories, Inc., 2001.
[6] V. Martinez, F. Reventós, and C. Pretel, “Post-test calculation of the ROSA/LSTF Test 3-1 using RELAP5/Mod3.3,” NUREG/IA-0409, USNRC, Washington, DC, 2012.
[7] S. Gallardo, V. Abella, G. Verdú, and A. Querol, “Assessment of TRACE 5.0 against ROSA Test 3-1, cold leg SBLOCA,” NUREG/IA-0413, Washington, DC, 2012.
[8] N. Zuber, “Problems in Modeling Small Break LOCA,” NUREG-0724, USNRC, Washington, DC, 1980.
[9] V.G. Zimin, H. Asaka, Y. Anoda, and M. Enomoto, “Verification of J-TRAC code with 3D neutron kinetics model SKETCH-N for PWR rod ejection analysis,” in Proc. of the 9th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9), San Francisco, USA, October 1999.
[10] H.K. Fauske, “The discharge of saturated water through tubes,” AlChE Symp. Ser., vol. 61, 1965, pp. 210–216.
[11] K.H. Ardron and R.A. Furness, “A study of the critical flow models used in reactor blowdown analysis,” Nucl. Eng. Des., vol. 39, 1976, pp. 257–266.
[12] D.W. Sallet, “Thermal hydraulics of valves for nuclear applications,” Nucl. Sci. Eng., vol. 88, 1984, pp. 220–244.
[13] G.B. Wallis, “One-Dimensional Two-Phase Flow,” McGraw-Hill Book, New York, USA, 1969.
[14] T. Yonomoto, Y. Anoda, Y. Kukita, and Y. Peng, “CCFL characteristics of PWR steam generator U-tubes,” in Proc. of the ANS International Topical Meeting on Safety of Thermal Reactors, American Nuclear Society, Portland, Ore, USA, July 1991.
[15] T. Takeda, A. Ohnuki, and H. Nishi, “RELAP5 code study of ROSA/LSTF experiments on PWR safety system using steam generator secondary-side depressurization,” J. Energy Power Eng., vol. 9, 2015, pp. 426–442.
[16] M. Ishii and K. Mishima, “Study of Two-fluid Model and Interfacial Area,” NUREG/CR-1873, Argonne National Laboratory, Lemont, IL, 1980.
[17] H. Kumamaru, Y. Kukita, H. Asaka, M. Wang, and E. Ohtani, “RELAP5/MOD3 code analyses of LSTF experiments on intentional primary-side depressurization following SBLOCAs with totally failed HPI,” Nucl. Technol., vol. 126, 1999, pp. 331–339.
[18] L.A. Bromley, “Heat transfer in stable film boiling,” Chem. Eng. Prog., vol. 46, 1950, pp. 221–227.
[19] K.H. Sun, J.M. Gonzalez-santalo, and C.L. Tien, “Calculations of combined radiation and convection heat transfer in rod bundles under emergency cooling conditions,” J. Heat Transfer, vol. 98, 1976, pp. 414–420.
[20] F.W. Dittus and L.M.K. Boelter, “Heat transfer in automobile radiators of the tubular type,” Int. Comm. Heat Mass Transfer, vol. 12, 1985, pp. 3–22.
[21] J.R. Sellars, M. Tribus, and J.S. Klein, “Heat transfer to laminar flows in a round tube or flat conduit: the Graetz problem extended,” Transactions of the ASME, vol. 78, 1956, pp. 441–448.
[22] S.W. Churchill and H.H.S. Chu, “Correlating equations for laminar and turbulent free convection from a vertical plate,” Int. J. Heat Mass Transfer, vol. 18, 1975, pp. 1323–1329.
[23] A. Guba, M. Makai, and L. Pál, “Statistical aspects of best estimate method–I,” Reliability Eng. Syst. Safety, vol. 80, 2003, pp. 217–232.
[24] A. de Crécy, P. Bazin, H. Glaeser, et al., “Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme,” Nucl. Eng. Des., vol. 238, 2008, pp. 3561–3578.
[25] R.L. Iman and M.J. Schortencarier, “A FORTRAN program and user’s guide for the generation of Latin hypercube and random samples for use with computer models,” NUREG/CR-3624, USNRC, Washington, DC, 1984.
[26] A. Yamamoto, K. Kinoshita, T. Watanabe, and T. Endo, “Uncertainty quantification of LWR core characteristics using random sampling method,” Nucl. Sci. Eng., vol. 181, 2015, pp. 160–174.
[27] W.W. Daniel, “Spearman rank correlation coefficient,” Applied Nonparametric Statistics (second ed.), PWS-Kent Publishing, Boston, MA, 1990.
[28] J. Freixa, T.W. Kim, and A. Manera, “Post-test thermal-hydraulic analysis of two intermediate LOCA tests at the ROSA facility including uncertainty evaluation,” Nucl. Eng. Des., vol. 264, 2013, pp. 153–160.